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Okubo, Nariaki; Sokolov, M. A.*; Tanigawa, Hiroyasu; Hirose, Takanori; Jitsukawa, Shiro; Sawai, Tomotsugu; Odette, G. R.*; Stoller, R. E.*
Journal of Nuclear Materials, 417(1-3), p.112 - 114, 2011/10
Times Cited Count：11 Percentile：60.18(Materials Science, Multidisciplinary)Irradiation hardening and fracture toughness of reduced-activation ferritic/martensitic steel F82H after irradiation were investigated with a focus on changing the fracture toughness transition temperature as a result of several heat treatments. The specimens were standard F82H-IEA (IEA), F82H-IEA with several heat treatments (Mod1 series) and a higher tantalum containing (0.1%) heat of F82H (Mod3). The specimens were irradiated up to 18 dpa at 300 
C in High Flux Isotope Reactor under a collaborative research program between JAEA/US-DOE. The results of hardness tests showed that irradiation hardening of IEA was comparable with that of Mod3. However, the fracture toughness transition temperature of Mod3 was lower than that of IEA. The transition temperature of Mod1 was also lower than that of the IEA heat. These results suggest that tightening of specifications on the heat treatment condition and modification of the minor alloying elements seem to be effective to reduce the fracture toughness transition temperature after irradiation.
Oyaizu, Makoto; Isobe, Kanetsugu; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1143 - 1146, 2011/10
Times Cited Count：7 Percentile：45.35(Materials Science, Multidisciplinary)The tritium permeation behavior through F82H steel from carrier-free tritiated water vapor was investigated and the difference between as-received sample surface and that after the permeation experiments were analyzed by means of SEM, EDX and XRD. As the results, the permeability of tritium through F82H steel from tritiated water vapor is two to three orders of magnitude smaller than that from gaseous hydrogen. A thick and porous iron oxide layer composed of hematite and magnetite was formed in the permeation experiments. These results indicate that the oxide layer would hardly work as tritium permeation barrier, and that gaseous hydrogen could be generated by the redox reaction of water on the sample surface. Therefore, it could be considered that the tritium permeation through F82H steel from tritiated water vapor could result from the partial pressure of T
.
Nishitani, Takeo; Tanigawa, Hiroyasu; Nozawa, Takashi; Jitsukawa, Shiro; Nakamichi, Masaru; Hoshino, Tsuyoshi; Yamanishi, Toshihiko; Baluc, N.*; M
slang, A.*; Lindou, R.*; et al.
Journal of Nuclear Materials, 417(1-3), p.1331 - 1335, 2011/10
Times Cited Count：14 Percentile：68.24(Materials Science, Multidisciplinary)As a part of the Broader Approach (BA) activities, the research and development on blanket related materials and tritium technology have been initiated toward DEMO by Japan and EU. Recently, those five R&D items have progressed substantially in Japan and EU. As a preparatory work aiming at the RAFM steel muss-production development, a 5-ton heat of RAFM steel (F82H) was procured with the Electro Slag Re-melting as a secondary melting. The result of the double notch tensile test method for the NITE-SiC
/SiC specimen indicated notch insensitivity and very minor size effect on proportional limit tensile stress and fracture strength. For the fabrication technology development of beryllide neutron multiplayer pebbles, Be- Ti inter-metallic pebbles have been sintered directly from the mixed powder of Be and Ti in Japan.
Nozawa, Takashi; Tanigawa, Hiroyasu
Journal of Nuclear Materials, 417(1-3), p.440 - 444, 2011/10
Times Cited Count：5 Percentile：35.49(Materials Science, Multidisciplinary)Development of small specimen test technique (SSTT) is one of key issues for irradiation study of nuclear-grade materials. This study aims to propose a new tensile test methodology using miniature notched specimens. For that purpose, crack extension behaviors of notched tensile specimens and their size effects were initially evaluated. Observing crack propagation behaviors of SiC/SiC composites, the apparent notch insensitivity was identified even if the moderate fiber/matrix interface was formed. Due to the notch insensitivity, it was found that key tensile properties can easily be predicted by test results of notched specimens. Noteworthy, this relation appeared independent of specimen size. Based on these findings, a conceptual design of miniature tensile test methodology using a notched specimen was proposed.
Nakazawa, Tetsuya; Igarashi, Takahiro; Tsuru, Tomohito; Kaji, Yoshiyuki; Jitsukawa, Shiro
Journal of Nuclear Materials, 417(1-3), p.1090 - 1093, 2011/10
Times Cited Count：5 Percentile：35.49(Materials Science, Multidisciplinary)It is well known that impurities in iron which segregate to grain boundaries can dramatically change physical and chemical properties. Phosphorous, which is one of impurities, segregates at grain boundaries under thermal or irradiation environments, and brings about the intergranual embrittlement. In this study, influences of phosphorus substitutions for binding energies and electronic structures of octahedral iron cluster are investigated computationally using density functional calculations in order to understand the nature of bonding between phosphorus and iron at grain boundaries. The values of binding energies of cluster are increasing with the phosphorous substitutions. The increases are due to Fe-P bond strengthened by the charge transfer from Fe atom to P atom. On the other hand the calculated bond orders give that Fe-Fe bonds are weakened. Thus, the embrittlement induced with the segregation of P due to irradiation is considered to be associated with weakened Fe-Fe bonds.
Iwai, Yasunori; Kobayashi, Kazuhiro; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1187 - 1190, 2011/10
Times Cited Count：2 Percentile：16.92(Materials Science, Multidisciplinary)The concept of tritium containment and confinement is the root of fusion safety. Hence, HTO contamination on concretes and epoxy paint should be reduced as low as possible. Several kinds of hydrophobic coatings, a commercial silicic paint, a commercial acrylic paint, a commercial fluoric paint, methoxytrimethylsilane paint or metallic stick-sheets, were tested on concrete and epoxy samples. These samples were exposed to 740-1110Bq/cm
of HTO vapor at room temperature for a given period from 1 to 60 weeks. Static leaching tests were carried out for every HTO absorbed sample in distilled water, and the amount of leached HTO was evaluated. The hydrophobic barriers were effective to reduce HTO penetration into concrete. After exposure to HTO for 1 week, the HTO amount penetrated into concrete was reduced to 54.2% of non-paint sample for methoxytrimethylsilane paint, 56.0% for a commercial fluoric paint, 66.8% for a commercial silicic paint, respectively. Effectiveness of these hydrophobic barriers became less as the samples were exposed to HTO for a longer period.
Hirose, Takanori; Okubo, Nariaki; Tanigawa, Hiroyasu; Ando, Masami; Sokolov, M. A.*; Stoller, R. E.*; Odette, G. R.*
Journal of Nuclear Materials, 417(1-3), p.108 - 111, 2011/10
Times Cited Count：19 Percentile：77.12(Materials Science, Multidisciplinary)This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels irradiated at 573 K. The materials used in this research were F82H-IEA and its modified steels. Post irradiation mechanical tests revealed that irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated proof stress is less than 1 GPa. The deterioration of total elongation was also saturated by 9 dpa. Irradiation response of F82H-mod3, which is stable to temperature instability during material production and HIP treatment, was very similar to that of F82H-IEA, and negative impacts of extra tantalum was not observed. Therefore it can be an attractive option for the structural materials for blanket components manufactured by HIP.
Li enrichment using an ionic-liquid impregnated organic membraneHoshino, Tsuyoshi; Terai, Takayuki*
Journal of Nuclear Materials, 417(1-3), p.696 - 699, 2011/10
Times Cited Count：48 Percentile：94.21(Materials Science, Multidisciplinary)The tritium as a fuel for fusion reactors is produced by the reaction of lithium-6 (Li) with a neutron in tritium breeding material. However, natural Li contains only about 7.6% Li, and the enrichment of Li up to 30 - 90% is required for tritium breeding material in the fusion reactor. The mercury amalgam method is superior as one of the lithium isotope enrichment methods, and might be currently utilized in practice. In Japan, on the other hand, the development of lithium isotope enrichment methods using the ion exchange membrane and molten salt has been conducted to avoid the environmental pollution. However, the isotope separation coefficient and efficiency is too low in the case of these methods. Therefore, these methods were difficult to apply to mass production for the large need of fusion blanket. Preliminary experiments were conducted. Organic membranes impregnated with TMPA-TFSI and PP13-TFSI as ionic liquids were prepared, and the relationship between the Li separation coefficient and the applied dialysis electric current was measured. The results showed that Li isotope separation coefficient by this method (about 1.2
1.4) was larger than that by the mercury amalgam method (about 1.06).
Nogiwa, Kimihiro; Nishimura, Akihiko; Yokoyama, Atsushi; Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Okubo, Tadakatsu*; Hono, Kazuhiro*
Journal of Nuclear Materials, 417(1-3), p.201 - 204, 2011/10
Times Cited Count：10 Percentile：57.03(Materials Science, Multidisciplinary)Du se 9Cr-ODS (oxide dispersion-strengthened) steel consisting of residual-
ferrite and prime martensite has excellent high-temperature strength. This study describes the microstructure of dual-phase 9Cr-ODS steels characterized by atom-probe tomography in order to compare oxide-particle dispersion states in each phase. This revealed that nano-size oxide particles were of the same chemical composition and that their mean size was about 3 nm in each phase. On the other hand, the number density in the residual- phase was about four times higher than that of the prime phase. These results indicate that the dense distribution of the oxide particles in the residual- phase contribute to the excellent high-temperature strength of 9Cr-ODS steel.
Nakamichi, Masaru; Yonehara, Kazuo
Journal of Nuclear Materials, 417(1-3), p.765 - 768, 2011/10
Times Cited Count：47 Percentile：93.99(Materials Science, Multidisciplinary)-FeSuzudo, Tomoaki; Kaburaki, Hideo; Yamaguchi, Masatake
Journal of Nuclear Materials, 417(1-3), p.1102 - 1105, 2011/10
Times Cited Count：12 Percentile：62.85(Materials Science, Multidisciplinary)This paper discusses a computational approach to the helium segregation at grain boundaries with reference to the helium embrittlement under the fast neutron irradiation; we particularly applied a kinetic Monte-Carlo method to the modeling of the segregation phenomena in alpha-iron. In the calculation, the migration and segregation energies of helium atoms given by the first principle calculations are used. Obtained results are discussed in comparison with McLean's equilibrium segregation theory.
Nakamura, Hirofumi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1150 - 1153, 2011/10
Times Cited Count：3 Percentile：23.83(Materials Science, Multidisciplinary)Deuterium permeation characteristics through various surface states ofF82H such as F82H without surface treatment (bare F82H), welded F82H,and gold plated F82H (Au-F82H) have been investigated in order to understand the effect of surface state on permeation for the realistic tritium permeation evaluation in the fusion reactors especially breeding blanket system, which has many welding points and permeation reduction coatings. Based on obtained permeation behavior, the steady state permeation and the diffusivity derived form the transient permeation behavior have been discussed. As the results, deuterium permeation through bare F82H is smaller than that through clean surface F82H. Asto the effect of welding on permeation, the results indicate that significant difference of welding is not observed between bare F82H and welded F82H. Finally, gold plating on F82H showed good permeation reduction performance.
Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1183 - 1186, 2011/10
Times Cited Count：2 Percentile：16.92(Materials Science, Multidisciplinary)In a fusion reactor of high safety and acceptability, safe confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facility of ITER will be used various construction materials such as the concrete, the organic materials. As the results, the concrete materials were almost saturated with HTO vapor within about 1month except for cement paste and it was larger in the order of cement paste 
mortar concrete. Even if one month passes from the exposure beginning, the amount of sorbed tritium to cement paste did not reach saturation. The chemical form of desorbed tritium from the sample was almost HTO. In addition, the tritium behavior that adsorbs the surface of concrete materials will be discussed by using FT-IR.
Sato, Satoshi; Maegawa, Toshio*; Yoshimatsu, Kenji*; Sato, Koichi*; Nonaka, Akira*; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara
Journal of Nuclear Materials, 417(1-3), p.1131 - 1134, 2011/10
Times Cited Count：12 Percentile：62.85(Materials Science, Multidisciplinary)The multi-layered concrete structure has been developed to reduce induced activity in the concrete applied for neutron generation facilities such as a fusion reactor. The multi-layered concrete structure is composed of the low activation concrete as the first layer, the boron-doped low activation concrete as the second layer and the ordinary concrete as the third layer from the side of the neutron source. By applying the multi-layered concrete structure, the volume of the boron can drastically decrease compared with the monolithic boron-doped concrete. A 14 MeV neutron irradiation experiment with the multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activities in the mockups were measured. This experiment demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activities.
Sukegawa, Atsuhiko; Anayama, Yoshimasa*; Okuno, Koichi*; Sakurai, Shinji; Kaminaga, Atsushi
Journal of Nuclear Materials, 417(1-3), p.850 - 853, 2011/10
Times Cited Count：25 Percentile：83.80(Materials Science, Multidisciplinary)A flexible heat resistant neutron shielding material has been developed, which consists of polymer resin with 1 weight % boron. The neutron shielding performance of the developed resin, examined by the 
Cf neutron source is almost the same as that of the polyethylene. The outgas of H, HO, CO and CO from the resin have been measured at 250 C environment. The resin will be applied around the port of the vacuum vessel as an additional shielding material and prevented the effects on the neutron streaming of the superconducting tokamak device such as JT-60SA.
Taguchi, Tomitsugu; Hasegawa, Yoshio*; Shamoto, Shinichi
Journal of Nuclear Materials, 417(1-3), p.348 - 352, 2011/10
The SiC/SiC composites with and without dispersed carbon nanofiber were fabricated by polymer impregnation and pyrolysis process. The effect of dispersing with 6 vol.% carbon nanofiber on the mechanical and thermal properties of SiC/SiC composites was investigated. The bending strength and elastic modulus of SiC/SiC composites with carbon nanofiber decreased slightly compared to those of the SiC/SiC composites without carbon nanofiber. On the other hand, the thermal conductivity of SiC/SiC composites increased with increasing the amount of dispersed carbon nanofiber. The dominant reason is considered that the pore shape changed from oblong shape vertical to direction of heat flow to isotropic shape by dispersing carbon nanofiber.
Alimov, V.; Shu, Wataru*; Roth, J.*; Lindig, S.*; Balden, M.*; Isobe, Kanetsugu; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.572 - 575, 2011/10
Times Cited Count：96 Percentile：98.66(Materials Science, Multidisciplinary)Blistering and deuterium retention in re-crystallized tungsten exposed to a low energy (38 eV/D) and high deuterium ion flux (10
D/m
s) D plasma at ion fluences of 10
and 10
D/m at temperatures in the range from 320 to 800 K have been examined with scanning electron microscopy, thermal desorption spectroscopy (TDS), and the nuclear reaction. During exposure to the D plasma blisters with various shapes and sizes depending on the exposure temperature are formed on the W surface. At the temperatures above 700 K the blisters disappear. The deuterium retention increases with the exposure temperature, reaching its maximum value of about 7
10
D/m at 530 K and about 110 D m at 480 K for ion fluences of 10 and 10 D/m, respectively. As the temperature grows further, the D retention decreases to about 10
D/m at 800 K.
TiO
during long-term irradiation in the WWR-K reactorTazhibayeva, I.*; Beckman, I.*; Shestakov, V.*; Kulsartov, T.*; Chikhray, E.*; Kenzhin, E.*; Kuykabaeva, A.*; Kawamura, Hiroshi; Tsuchiya, Kunihiko
Journal of Nuclear Materials, 417(1-3), p.748 - 752, 2011/10
Times Cited Count：17 Percentile：74.02(Materials Science, Multidisciplinary)For the first time the data was obtained on tritium release from Li-enriched (96%) lithium metatitanate under high lithium burn-up (up to 23%). Proposed mathematics and software of the reactor experiments allowed to interpret the experimental results of tritium release study. Tritium was continuously generated as a result of the nuclear reaction of lithium-6 and thermal neutrons under variable thermal impacts (graduated heating and cooling) on lithium metatitanate LiTiO. Main gas release parameters were calculated in order to assess acceptability of the use of lithium metatitanate granules in tritium breeders; the parameters are as follows: gas release rate, tritium retention in the materials, retention time, activation energy of thermal desorption HT, activation energy of volume diffusion T
, as well as corresponding pre-exponential (frequency) indexes. It was discovered that the tritium release process is mainly controlled by tritium volume diffusion, however, capture of tritium by the point defects and tritium molization at the material's surface played the certain role in the process as well. It was discovered that as lithium is burnt-up, the activation energy of tritium release decreases and tends to a constant value under high lithium-6 burn-up.
Hoshino, Tsuyoshi; Kato, Kenichi*; Natori, Yuri*; Oikawa, Fumiaki; Nakano, Natsuko*; Nakamura, Mutsumi*; Sasaki, Kazuya*; Suzuki, Akihiro*; Terai, Takayuki*; Tatenuma, Katsuyoshi*
Journal of Nuclear Materials, 417(1-3), p.684 - 687, 2011/10
Times Cited Count：55 Percentile：95.48(Materials Science, Multidisciplinary)Lithium titanate (LiTiO) is one of the most promising candidates among tritium breeding materials because of its good tritium release. Addition of H to inert sweep gas has been proposed for enhancing the tritium release from tritium breeding materials. However, the mass of LiTiO was decreased with time in the hydrogen atmosphere. It is assumed that the mass decrease indicates the loss of the oxygen contained in the sample caused by the change from Ti 
to Ti 
, and that the partial pressures of Li-containing species were increased in the hydrogen atmosphere. In order to decrease the mass-change at high temperature, advanced tritium breeding material with added Li should be developed to improve the physical and chemical stability in hydrogen atmosphere. In the case of the Li
TiO
samples used by the present study, LiOHHO and HTiO were proportionally mixed with the molar ratio Li/Ti of either 2.0 and 2.2. These samples are designated as L20 (Li/Ti = 2.0) and L22 (Li/Ti = 2.2), respectively. The results of XRD measurement showed that the phases in advanced tritium breeding material were as follows. L22 existed as non-stoichiometric compound LiTiO.
Vayakis, G.*; Bertalot, L.*; Encheva, A.*; Walker, C.*; Brichard, B.*; Cheon, M. S.*; Chitarin, G.*; Hodgson, E.*; Ingesson, C.*; Ishikawa, Masao; et al.
Journal of Nuclear Materials, 417(1-3), p.780 - 786, 2011/10
Times Cited Count：31 Percentile：88.21(Materials Science, Multidisciplinary)